In a nuclear plant of the pressurized water reactor (PWR) type, coolant fluid, which is basically boron and water, is continuously transferred through a closed circulation loop between a nuclear reactor and one or more steam generators.
During power production, the pressurized coolant absorbs heat released by the thermonuclear reaction occurring in the reactor. The heated coolant then flows through a main pipe which is appropriately known as the "hot leg" of the circulation loop. The hot leg delivers the hot coolant to a steam generator.
In the steam generator, the coolant fluid circulates through a heat exchanger. The heat exchanger cools the coolant fluid and uses the heat removed from the coolant to produce steam. This steam is eventually used to drive turbines and generate electricity.
After the circulating coolant is cooled by a heat exchanger, a circulation pump removes the coolant from the steam generator via a "suction leg" and returns it to the reactor via a "cold leg" and inlet. The coolant is then reheated in the reactor and the cycle repeats.
This circulation of coolant through one or more loops is critical for the operation of the power plant. Not only does it deliver heat energy to the steam generators where the energy is used to produce steam for driving the turbines, but the circulating coolant also prevents the reactor core in the reactor from over-heating.
Nuclear power plant systems, including the steam generators, require periodic maintenance. In particular, the fluid circulation system must be inspected for potential degradation and nozzle dams must be installed and removed from the steam generators to allow inspection and maintenance to be performed in a dry environment.
In order to install and remove nozzle dams, the coolant fluid must be drained from the steam generator. This requires lowering the fluid level in the main circulation loop and consequently the hot leg or main pipe. During such a maintenance period, which is termed a "shutdown", the coolant continues to be heated by decay heat from the reactor core but it is cooled by an alternate heat exchanger and auxiliary circulatory system know as the "shutdown cooling system".
In order to lower the coolant or water level in the shutdown reactor system to allow maintenance operations on portions of the system above the lowered water level, the water level must be controlled and maintained at a minimum level and flow rate to continuously provide adequate core cooling. This minimum level is about midway within the reactor coolant system main loop piping (the hot leg) and is commonly referred to as "midloop".
During midloop operation, coolant water is circulated through the system to cool the core. Typically, there is a drain line or lines which communicate with the lower region of one or more of the main loop pipes or legs to draw the heated water from the core for cooling by the alternate heat exchanger in the shutdown cooling system and subsequent recirculation of cooled water to a reactor inlet and thus to the core.
It is possible to experience the formation of a coriolis effect vortex in the drain line during midloop operation if the water level is lowered too far down or if the drain flow rate is too high. Such a vortex is undesirable because it limits the rate at which coolant flow can be drained from the system and it can eventually lead to cavitation in the drain pump. Both results cause concern for continued cooling of the core.
The current methods to avoid vortex formation attempt to keep the water level as high as possible and/or reduce the flow rate, resulting in a conflict between the need to lower the water level for maintenance service, and the need to keep the water level high and at a sufficient rate for safe core cooling.
Midloop measuring systems in use are related to a detection of the water elevation and inference of the status of the vortex therefrom. Co-pending U.S. patent application Ser. No. 08/783,978, assigned to the same assignee as the instant application and filed Jan. 15, 1997, is for a system which reduces pump cavitation by inserting a vortex breaker in the drain pipe adjacent the main pipe or hot leg.
In nuclear power plants much attention has been given to shutdown cooling system reliability, especially during reactor coolant system midloop water level operation. Midloop operation in a typical pressurized water reactor (PWR) nuclear steam supply system, for example, for the installation and removal of steam generator nozzle dams, can be a very difficult operational process. In fact, typically, the water level allowed tolerance is approximately plus or minus one inch (.+-.1"). The plant operator must control this water level manually. Current instruments used measure only the average reactor coolant system (RCS) water level and the shutdown cooling pump current.
The RCS water level measurement accuracy is limited by the instrument technology used and process parameter changes such as temperature, pressure and boric acid concentration. The shutdown cooling pump current measurement alarm occurs only after air has already been ingested into the pump, thus it cannot be used to avoid the air vortex.